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Local neutral pressure measurements in the closed small angle slot (SAS) divertor [1,2] on DIII-D show a large increase when the divertor plasma shifts from high recycling into detachment. In-tile pressure gauges were installed to measure the pressure in the near- and far-SOL regions to examine the predicted high neutral pressures and compression. Cross-field drift effects lead to ∼ 10 × higher peak neutral pressure in detachment with the ion ∇B drift toward the divertor compared to out of the divertor, with similar pressures in attached conditions. Drifts also play a role in the neutral distribution in the slot while attached but become less pronounced in detachment once gradient drives reduce. Variation in the outer strike point location found higher neutral pressure and detachment at modestly lower main-plasma density with the strike point positioned away from the designed operation point. Reducing the fraction of the SOL width allowed into the slot increases neutral leakage into the main chamber and increases the main-plasma density required for detachment. Preliminary modeling with the SOLPS code without drifts over-predicts the neutral pressure in detachment by a factor of 2 with the strike point in the designed operation point while more significantly over-predicts the neutral slot compression; experiments show a broader distribution of neutrals through the slot. These measurements are used to help understand detachment and validate divertor design metrics. Keywords: DIII-D, Divertor, Neutral pressure, Detachment

Actively enhancing plasma radiated power has been used to reduce divertor heat flux in high power, high-performance double null divertor (DND) and near-DND plasmas in DIII-D, while at the same time maintaining acceptable energy confinement. One form of this approach that we focus on here is the radiating mantle. The effectiveness of the radiating mantle was found to depend strongly on the location of electron cyclotron (EC) heating deposition and on the effect that seed impurities have on triggering inimical MHD activity inside the plasma. In addition, we present experimental support for the predictions made by the ELITE boundary stability code (Snyder et al., 2002, [1]) for ways to improve confinement and fueling in these high-performance plasmas. Representative AT operating conditions were H98 ≈ 1.4–1.7, βN ≈ 3–4, and q95 ≈ 4.7–6.5. JEL classification: Divertor Materials (D0500), Impurities (10100), Plasma Properties (P0600), Keywords: DIII-D, Edge pedestal, Energy confinement, Gas injection and fueling impurity, Power deposition

Experiments at DIII-D use different divertor geometries to measure the effect of divertor closure on detachment onset. We compare matched H-mode discharges in the closed (upper) divertor and the open (lower) divertor and use as a detachment onset metric the rollover in peak ion flux to the target, Jsat, as a function of upstream pedestal top density, ne,ped. Measurements of electron temperature in the lower divertor confirm that this metric captures the transition from attached to detached divertor for this data set. Experiments at two different neutral beam injection powers demonstrate an ≈ 20–40% decrease in pedestal density at detachment onset in the closed case compared to the open. High-resolution edge density and temperature measurements and power balance analysis allow us to determine the difference in upstream separatrix density ne,sep at detachment onset, which we find to be ≈ 15–25% lower in the closed divertor case. The effect of closure on detachment seen in separatrix density is hence only about 60–75% the effect seen in pedestal density, with the remainder due to differences in ne,sep/ne,ped. Future devices are expected to have a scrape off layer that is opaque to neutrals, and hence will not have the same pedestal fueling to affect ne,sep/ne,ped; this suggests that the effect of closure on detachment onset seen in future devices will be less than that seen in current experiments, and closer to the magnitude of the effect we see here in the separatrix for a similar degree of divertor closure. Keywords: Tokamak, Divertor, Detachment, Fusion

A refined version of the Fundamenksi-Moulton 'free-streaming' model (FSM) for the dynamics of divertor density, particle flux, and heat flux during edge localized modes (ELMs) is presented. This model depends only on inter-ELM pedestal and divertor conditions and, crucially, incorporates particle recycling: a FSM with recycling model, FSRM. The effective particle recycling coefficient, Reff, is the only empirical fitting parameter in the FSRM. The predictions of the FSRM are systematically tested against a DIII-D database of ELM ion and energy fluence measurements and are shown to be consistent with the model across a wide range of pedestal and divertor conditions using a constant value of 0.96 for Reff . Predictions for W sputtering during ELMs are developed based on the FSRM. It is concluded that energetic free-streaming D+ ions and C6+ impurities are the dominant contributors to the intra-ELM gross erosion of W in the DIII-D divertor, i.e., recycling ions and impurities have relatively little impact on the total W sputtering rate. These calculations are also shown to be consistent with spectroscopic measurements of W gross erosion for three different pedestal conditions after incorporating the strong electron density dependence of the WI 400.8 nm ionizations/photon (S/XB) coefficient. Keywords: Tungsten sputtering, Tungsten Erosion, WI spectroscopy, Edge Localized Modes, Recycling

The structure of the edge plasma in a magnetic confinement system has a strong impact on the overall plasma performance. We uncover for the first time a magnetic-field-direction dependent density shelf, i.e., local flattening of the density radial profile near the magnetic separatrix, in high confinement plasmas with low edge collisionality in the DIII-D tokamak. The density shelf is correlated with a doubly peaked density profile near the divertor target plate, which tends to occur for operation with the ion $B\ifmmode\times\else\texttimes\fi{}\ensuremath{\nabla}B$ drift direction away from the $X$-point, as currently employed for DIII-D advanced tokamak scenarios. This double-peaked divertor plasma profile is connected via the $E\ifmmode\times\else\texttimes\fi{}B$ drifts, arising from a strong radial electric field induced by the radial electron temperature gradient near the divertor target. The drifts lead to the reversal of the poloidal flow above the divertor target, resulting in the formation of the density shelf. The edge density shelf can be further enhanced at higher heating power, preventing large, periodic bursts of the plasma, i.e., edge-localized modes, in the edge region, consistent with ideal magnetohydrodynamics calculations.

An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based on experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. These techniques may find applications in the future devices.

We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth) and realistic (rough) surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms. Keywords: Erosion, Deposition, Migration, Roughness, Aluminum, Mixed materials

As High-Z materials will likely be used as plasma-facing components (PFCs) in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13 C deposition resulting from local 13 CH 4 injection also provides information on radial transport due to E × B drifts and cross field diffusion. Finally, D 2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.

The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss power to the l -H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection with up to 50% of the injected power radiating in the core cools the pedestal and core plasmas, thus significantly degrading the confinement. As for Ne seeding, medium confinement with H98∼0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N2 seeded plasmas, radiation is predominately confined in the boundary plasma, which leads to less effect on the confinement and pedestal. In the case of strong N2 gas puffing, the confinement recovers during the detachment, from ∼20% reduction at the onset of the detachment to greater than unity comparable to that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.

Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND) hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P) ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5) due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot”) but also that heating near the slot opening is a significant source for impurity contamination of the core. PSI-22 keywords: DIII-D, Divertor geometry, Gas injection and fueling, Impurity sources, Power deposition

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower $n_{e}$ , and ELM energy and divertor peak heat flux reduction, especially prominent in radiative $D_{2}$ -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative $D_{2}$ -seeded SF divertor at $P_{\text {SOL}}=3-4$ MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected $P_{\text {SOL}} \simeq 9$ MW case. The radiative SF divertor with carbon impurity provides a wider $n_{e}$ operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar $q_{\text {peak}}$ reduction factors (see standard divertor).

A new small angle slot (SAS) divertor concept has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Initial tests on DIII-D find a strong interplay between such anticipated 'SAS' effects and cross-field drifts, favouring operation with the ion B × ∇B drift away from the X-point, as currently employed for advanced tokamaks. This offers the following key improvements relative to DIII-D's open lower divertor or partially-closed upper divertor: (i) SAS allows for transition to low temperature moderately detached divertor conditions with T e ≲ 10 eV at very low main plasma densities, lower than are usually attainable at all in DIII-D high confinement (H-mode) plasmas as used in these tests; (ii) Pedestal performance and core confinement are significantly improved with SAS. The final confinement collapse associated with the onset of X-point MARFE (multifaceted asymmetric radiation from the edge) following deep detachment occurs at significantly higher pedestal densities, thus widening the window of H-mode operation compatible with a dissipative divertor. For operation with the ion B × ∇B drift toward the X-point, the divertor plasma transitions to a bifurcative detached state at much higher densities, similar to other divertor configurations in DIII-D. These results highlight the strong interplay between divertor closure and drifts, and point to an interesting divertor optimization path to explore that offers potential for future fusion reactors.

Feedback control of radiated power from the lower divertor P rad, div, L has been implemented in the DIII-D Plasma Control System (PCS). A realtime sensor for P rad, div, L has been constructed from 12 foil bolometer channels which agrees with standard post-shot analysis to within 20%. Results with the 12-channel sensor are compared to initial proof-of-concept tests with a single channel as a proxy for P rad, div, L , showing that the upgraded sensor is necessary to overcome limitations of the proxy channel strategy in DIII-D. Using N 2 seeding under feedback control, P rad, div, L has been increased by up to 150% above unseeded levels, and a radiated power fraction f rad of 80% has been demonstrated, although feedback controlled gas flow is steadier at f rad = 55%. Spatial coverage is broad enough to enable P rad control during the strike point sweeps which are commonly used to generate pseudo-2D divertor Thomson measurements in DIII-D divertor experiments. Use of this control reveals challenges that may affect next step devices, which will require actively controlled extrinsic impurity seeding in order to manage heat loads. When operating at high f rad , changes in pedestal T e (caused by ELMs in these experiments but could come from other disturbances) resulted in large perturbations to P rad which were destabilizing to the feedback controller.

The compatibility of detached radiative divertor operation with a robust H-mode pedestal is examined in DIII-D. A density scan produced low temperature plasmas at the divertor target, Te ⩽ 2 eV, with high radiation leading to a factor of ⩾4 drop in peak divertor heat flux. The cold radiative plasma was confined to the divertor and did not extend across the separatrix in X-point region. A robust H-mode pedestal was maintained with a small degradation in pedestal pressure at the highest densities. The response of the pedestal pressure to increasing density is reproduced by the EPED pedestal model. However, agreement of the EPED model with experiment at high density requires an assumption of reduced diamagnetic stabilization of edge Peeling–Ballooning modes.

Aluminum (Al) is a convenient proxy for beryllium (Be) plasma material interaction studies since they have a number of physical and chemical similarities. Al samples were exposed at the lower outer strike point of an L-mode divertor plasma in DIII-D (conditions 7–11 × 1018 D-ions cm−2 s−1, Te = 12–47 eV). The gross erosion rate was directly measured using post-mortem ion beam analysis of small 1 mm-sized samples where local re-deposition was determined to be negligible. The gross erosion rate was also calculated using spectroscopic methods, but these rates greatly underestimate the direct (i.e. non-spectroscopic) measurement. The direct measured erosion yields were within the range of published D+ → Al ion beam sputtering yields. The ionizations per photon (S/XB) coefficients used in the spectroscopic analysis were determined in separate experiments using He plasmas at the PISCES-B linear plasma facility at UCSD. The measured S/XB coefficients were on average ∼6× higher than the theoretically calculated values.

We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux ( q ⊥ p ) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β N ≅ 3.0 and H 98(y,2) ≈ 1.35. It is also demonstrated that q ⊥ p could be reduced ≈50% by extending the parallel connection length ( L ||-XPT ) in the scrape-off layer between the X -point and divertor targets over a variety of the RD and non-RD environments tested.

The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for D α , D β and D γ for values of T e and n e characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to T e .

The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.

A study to examine some current issues in the physics of the plasma sheath has been recently carried out in DIII-D low power Ohmic plasmas using both flush and domed Langmuir probes, divertor Thomson scattering (DTS), an infrared camera (IRTV), and a new calorimeter triple probe assembly mounted on the Divertor Materials Evaluation System (DIMES). The sheath power transmission factor was found to be consistent with the theoretically predicted value of 7 (±2) for low power plasmas. Using this factor, the three heat flux profiles derived from the LP, DTS, and calorimeter diagnostic measurements agree. Comparison of flush and domed Langmuir probes and divertor Thomson scattering indicates that proper interpretation of flush probe data to get target plate density and temperature is feasible and could potentially yield accurate measurements of target plate conditions where the probes are located.

Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D 2 D 2 seeding. The SF configuration was maintained for many energy confinement times (2–3 s) in H-mode discharges ( I p =1.2 I p = 1.2 MA, P NBI =4–5 P NBI = 4 – 5 MW, and B×∇B B × ∇ B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary B p B p null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

Edge fluid–plasma/kinetic–neutral modeling of well-diagnosed DIII-D experiments is performed in order to document in detail how well certain aspects of experimental measurements are reproduced within the model as the transition to detachment is approached. Results indicate, that at high densities near detachment onset, the poloidal temperature profile produced in the simulations agrees well with that measured in experiment. However, matching the heat flux in the model requires a significant increase in the radiated power compared to what is predicted using standard chemical sputtering rates. These results suggest that the model is adequate to predict the divertor temperature, provided that the discrepancy in radiated power level can be resolved.

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ⩾ 15 eV to ⩽3 eV occurring both at the outer strike point and upstream of the X-point. These observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

Simulations using the multi-fluid code UEDGE indicates that, in low confinement (L-mode) plasmas in DIII-D, poloidal projection of the ionization driven flows dominate poloidal particle flows in the divertor near the divertor plates, whereas E × B drift flows dominate the radial particle flows. In contrast, in high confinement (H-mode) conditions E × B drift flows dominate both poloidal and radial particle flows in the divertor in the vicinity of the strong gradient region near the separatrix. UEDGE indicates that the toroidal C2+ flow velocities in the divertor plasmas are mainly entrained within 30% to the background deuterium flow in both L- and H-mode plasmas in the plasma region where the CIII 465 nm emission is measured. Therefore, UEDGE indicates that the Doppler Coherence Imaging Spectroscopy (CIS), measuring the toroidal velocity of the C2+ ions, can provide insight to the deuterium flows in the divertor. Parallel-to-B velocity dominates the toroidal divertor flow; direct drift impact being less than 1%. Toroidal divertor flow is predicted to reverse when the magnetic field is reversed. This is explained by the parallel-B flow towards the nearest divertor plate corresponding to opposite toroidal directions in opposite toroidal field configurations.

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas. Keywords: SOL width, Radio-frequency heating, Neutral beam heating, Low collisionality

The OEDGE code is used to model core fueling for attached L-mode plasmas and between edge localized modes (ELMs) for attached H-mode plasmas in DIII-D. Empirical plasma reconstruction has been used to determine the plasma conditions in these discharges. EIRENE is used to model the hydrogen recycling. Divertor recycling accounts for 65–100% of the core fueling. The fraction of the total divertor target flux ionized inside the separatrix ranges from 5% to 20%. The fraction of total wall flux ionized inside the separatrix ranges from 20% to 50%. Neutrals originating from wall regions closer to the separatrix are more likely to ionize in the confined plasma. Ionization in the confined plasma is concentrated below the midplane with peaks in the poloidal profiles just above the X -point. Radial core ionization in high density H-mode is peaked strongly near the separatrix.

Comparisons have been made between heat flux measurements from Langmuir probes and embedded thermocouples in the divertor of DIII-D. Good agreement has been found near the outer strike point (OSP) during L-mode operation with Neutral Beam Injection (NBI) using a sheath power transmission factor (SPTF) of 7, predicted by collisionless 1-D sheath theory. Previous SPTF measurements taken from Langmuir probes and IR imagery on DIII-D demonstrated values below the theoretical limit. The Langmuir probe array has since been upgraded and an embedded thermocouple array has been utilized to measure heat flux. The SPTF has also been measured during a NBI heated H-mode shot. This shot demonstrated a SPTF greater than 7 neat the OSP, which is due to a larger scrape-off layer (SOL) current density during H-mode operation. These studies represent a significant advancement towards finding agreement between theoretical predictions of the SPTF at the divertor and experimental measurements from the divertor diagnostics.

This paper documents first work toward validation of BOUT++ nonlinear edge localized mode (ELM) simulations in X-point geometry, at experimental pedestal collisionality, against multiple diagnostic measurements of a well-characterized ELM event in DIII-D. The key to the BOUT++ simulations is the use of a hyper-resistivity model that effectively spreads the very thin current sheets that form in low collisionality nonlinear simulations, and allows for ELM driven magnetic reconnection at finite current density. Experimental ELM characterization includes multiple fast line-integrated diagnostic measurements revealing in–out divertor asymmetric response to ELMs, IRTV imaging at the divertor targets, visible emission in the divertor volume to test the extension of BOUT++ to X-point geometry, and forward modeling of new electron cyclotron emission imaging to test predictions of ELM filaments in the edge pedestal. Initial comparisons suggest optimized BOUT boundary conditions and model parameters, and show similarities between initial BOUT++ results and several measurements.

We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR ∝ [RTAR]2[L||]6/7 and TTAR ∝ [RTAR]−2[L||]−4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.

New scaled carbon atomic electron-impact excitation data is utilized to evaluate comparisons between experimental measurements and fluid emission modeling of detached plasmas at DIII-D. The C I and C II modeled emission lines for 909.8 and 514.7 nm were overestimated by a factor of 10–20 than observed experimentally for the inner leg, while the outer leg was within a factor of 2. Due to higher modeled emissions, a previous study using the UEDGE code predicted that a higher amount of carbon was required to achieve a detached outboard divertor plasma in L-mode at DIII-D. The line emission predicted by using the new scaled carbon data yields closer results when compared against experiment. We also compare modeling and measurements of D α emission from neutral deuterium against predictions from newly calculated R-Matrix with pseudostates data available at the ADAS database.

Experimental observation of net erosion of molybdenum being significantly reduced compared to gross erosion in the divertor of DIII-D is reported for well-controlled plasma conditions. For the first time, gross erosion rates were measured by both spectroscopic and non-spectroscopic methods. In one experiment a net erosion rate of 0.73 ± 0.03 nm/s was measured using ion beam analysis (IBA) of a 1 cm diameter Mo-coated sample. For a 1 mm diameter Mo sample exposed at the same time the net erosion rate was higher at 1.31 nm/s. For the small sample redeposition is expected to be negligible in comparison with the larger sample yielding a net to gross erosion estimate of 0.56 ± 12%. The gross rate was also measured spectroscopically (386 nm MoI line) giving 2.45 nm/s ± factor 2. The experiment was modeled with the REDEP/WBC erosion/redeposition code package coupled to the ITMC–DYN mixed-material code, with plasma conditions supplied by the OEDGE code using Langmuir probe data input. The code-calculated net/gross ratio is =0.46, in good agreement with experiment.

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

Thermo-oxidation is controlled exposure in anoxygen-containing atmosphere at elevated temperatureand is being considered as a technique for the de-tritiation of carbon-based codeposits in ITER. In addi-tion, unplanned oxidation may also occur duringaccidental air ingress. The impact of thermo-oxidationon ITER diagnostic mirrors causes concerns. A dedi-cated study was performed in DIII-D, where molybde-num and copper mirrors were installed in the mainchamber, in the divertor, and at a location remote fromthe plasma and exposed for

New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. 13 CH 4 was injected into the secondary outer divertor to simulate sputtering there. The majority of the 13 C found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall 13 CH 4 injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240 °C, is incomplete. The principal features of the measured 13 C deposition pattern have been replicated by the OEDGE interpretive code.

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. The SFD resulted in a 2.5× reduction in the peak heat flux for many energy confinement times (2–3 s) without any adverse effects on core plasma performance.

Fueling of the pedestal from divertor and main chamber recycling in DIII-D is assessed by a combination of experimental measurements, plasma modeling and kinetic neutral analysis. The 2D profile of surface recycling between edge localized modes in H-mode was previously determined in DIII-D from surface mounted Langmuir probes and a plasma configuration where toroidally symmetric recycling was dominant. The measured profile of recycling neutrals is launched by the DEGAS2 code into a 2D background plasma profile reconstructed by the UEDGE fluid model and constrained to match the upstream temperature and density profiles. Divertor plasmas of varying density and temperature are constructed to test sensitivity to uncertainty in the divertor plasma conditions. The pedestal ionization rate from this analysis can match the inferred outward ion flux across separatrix only for high density, detached inboard divertor conditions.

Recent DIII-D experiments show that both magnetic balance and particle drifts are important to understanding how argon impurities accumulate in balanced and unbalanced double-null plasmas during ‘puff and pump’ radiating divertor operation. Unbalanced double-null shapes, which are biased in the direction opposite to the ion B × ∇ B drift direction, have produced the best result to-date in terms of coupling a radiating divertor approach with an H-mode plasma. The proximity to balanced double-null that can be used for puff-and-pump may depend on the width of the heat flux profile in the scrape-off layer. A comparison of plasma behavior in the open lower divertor to that of the more closed upper divertor determined that plasma density control was far more sensitive to the ion B × ∇ B drift direction than to the relative closure of the divertor.

Flow velocities of the order 10-20 km/s in the direction of the high-field side divertor have been measured for deuterons and low charge-state carbon ions in the scrape-off layer at the crown of low-density L-mode plasmas, suggesting that these carbon ions at the crown move with the background plasma flow. Simulations with the multi-fluid edge code UEDGE including cross-field drifts due to E x B and B x del B yield calculated divertor conditions which are more consistent with the measurements, but flows at the crown that are stagnant or in the opposite direction than observed. The simulations indicate that both the ion temperature gradient force and deuteron frictional drag play a role in determining the flow direction and magnitude of low charge-state carbon ions. The effect of the assumed radial transport model, toroidal core rotation, and neutral pumping at the divertor plates on the flow at the crown is investigated. (C) 2009 Elsevier B.V. All rights reserved.

13 C methane puffing experiments were conducted on DIII-D in both L- and ELMy H-mode conditions. The puffing was toroidally symmetric into the crown of a series of well-characterized lower single-null discharges. The hydrocarbon breakup, carbon transport and deposition were modeled using the OEDGE interpretive code. Three separate hypotheses were tested using OEDGE to try to reproduce the experimental deposition: Radial variation of fast parallel flow, erosion of the puffed 13 C deposited in the divertor, and a pinch in the inner scrape off layer (SOL) towards the separatrix. A fast parallel flow was imposed for all hypotheses. The magnitude and the distribution of the 13 C deposition resulting from each hypothesis are compared. A fast parallel flow in the SOL toward the inner divertor combined with a pinch/drift of 10–30 m/s in the inner SOL towards the separatrix roughly reproduces the deposition in both the L- and H-mode experiments.

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.

Use of carbon in tokamaks leads to a serious tritium retention issue due to co-deposition. To further investigate the processes involved, a detached ELMy H-mode (6.5 MW NBI) experiment was performed on DIII-D in which {sup 13}CH{sub 4} was puffed into the main vessel through the toroidally symmetric pumping plenum at the top of lower single-null discharges. Subsequently, the {sup 13}C content of tiles taken from the vessel wall was measured. The interpretive OEDGE code was used to model the results. The {sup 13}C deposition pattern could be reproduced, in general shape and magnitude, by assuming in the code the existence of a parallel flow and a radial pinch in the scrape-off layer. Parallel flows of Mach {approx} 0.3 toward the inner divertor and a radial pinch {approx}10 to 20 m/s (+ R-direction) were found to yield {sup 13}C deposition comparable to the experiment.

A radiating divertor approach was successfully applied to high performance 'hybrid' plasmas [M.R. Wade, et al., Proc. 20th IAEA Fusion Energy Conf., Vilamoura, (2004)]. Our technique included: (1) injecting argon near the outer divertor target, (2) enhancing the plasma flow into the inner and outer divertors by a combination of particle pumping and deuterium gas puffing upstream of the divertor targets, and (3) isolating the inner divertor from the outer by a structure in the private flux region. Good hybrid conditions were maintained, as the peak heat flux at the outer divertor target was reduced by a factor of 2.5; the peak heat flux at the inner target decreased by 20%. This difference was caused by a higher concentration of argon at the outer target than at the inner target. Argon accumulation in the main plasma was modest (n{sub AR}/n{sub e} {le}0.004 on axis), although the argon profile was more peaked than the electron profile.

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration (D.D. Ryutov, Phys. Plasmas, 14, 064502, 2007) may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 – 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. The radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (cf. standard divertor).

Measurements and modeling of the 2D poloidal D α intensity distribution in DIII-D low density L-mode and medium density ELMy H-mode plasmas indicate that the core plasma is predominately fueled near the divertor X-point region. The neutral deuterium and ion carbon emission were measured in the divertor and inner main chamber scrape-off layer (SOL) using a plasma imaging technique, covering 85% of the poloidal cross-section. Typically, the peak emission in the inner main SOL at the tokamak midplane was three orders of magnitude lower than in the divertor. For discharges with the ion B × ∇ B drift direction toward the lower divertor the UEDGE/DEGAS codes predict strong core plasma fueling from the significantly higher density and lower temperature plasma calculated in the inner divertor leg. The concomitant carbon ion flow reversal in the inner divertor leg enhances the leakage of carbon from the divertor into the main SOL, and hence into the core.

Recent experiments on DIII-D point to the importance of two factors in determining the rate at which deuterium particles can be pumped at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of B x {del}B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the pumping rate at the inner divertor target(s) than on the pumping rate at the outer divertor target(s). The behavior in the particle pumping observed at the inner and outer divertor target(s) in the DN and near-DN shapes suggests a redistribution of particles that would be expected in the presence of B x {del}B and E x B ion particle drifts in the scrapeoff layer (SOL) and divertor(s).